SIMULATE5

SIMULATE5, Studsvik Scandpower’s next generation nodal code, has been benchmarked against recent gamma scan measurements of Oskarshamn2, an ABB/Westinghouse design BWR, and results are presented in this paper. Gamma scan measurements following the Cycle-32 operation includes 48 assembly gamma scans of advanced Optima2 and Atrium-10B fuels. Two Optima2 fuel assemblies were disassembled for more detailed gamma scanning of 49 fuel rods. SIMULATE5 is a three dimensional multi-group analytical nodal code with microscopic depletion capability. Its state of the art neutronic modules have been developed based on first principals with considerations of today’s highly heterogeneous cores. SIMULATE5 has a built in capability of tracking Ba140 isotopics in a node as well as fuel pin level which makes the evaluations of gamma scans straightforward. The assembly gamma scan comparisons reveal that SIMULATE5 accurately predicts the nodal Ba140 densities, hence nodal/assembly powers, regardless of fuel assembly type, burnup or location of the fuel assembly in the core. Heterogeneities in the axial direction, due to enrichment/burnable absorber zoning, part length fuel and spacer/grids are handled well. Pin gamma scan evaluations show that SIMULATE5 offers significant improvements in accuracy to calculate pin powers vs. existing methods, especially for the fuel assemblies with control rod which exhibits strong radial power/flux gradients. The last part of this paper summarizes the benchmark results for core tracking calculations for the last 18 cycles of Oskarshamn2.

SIMULATE5 is Studsvik’s next generation nodal code which has been developed along with CASMO5, Studsvik’s next generation lattice physics code, to address the deficiencies of existing reactor physics tools for today’s aggressive core designs with more heterogeneous fuel, extended cycle lengths, and operation strategies. This paper discusses the new neutronic and thermal-hydraulics models employed in SIMULATE5 as well as benchmarking activities against theoretical problems, critical experiments, core follow calculations, and gamma scan evaluations.

Reactor shutdown refueling is a crucial task for BWR operations. In this paper, based on a realistic BWR core refueling sequence, the SIMULATE3 and SIMULATE5 shutdown margin (SDM) predictions are benchmarked against CASMO5-M results. The 2D core in this exercise is established from the middle plane of a typical 748-bundle BWR core. There are thousands of fuel move steps from the end of previous cycle to the beginning of next cycle, most of which are so-called partially loaded core configurations where empty fuel locations are filled with water. During this process, eigenvalue predictions from CASMO5-M and SIMULATE3/5 are compared for the all-rod-in (ARI) case and the ARI except the maximum-worth rod (ARI-M) case. Based on more advanced neutronic models, SIMULATE5 shows sizable improvements in agreeing with the CASMO5-M reference solution over the SIMULATE3 results for partially loaded cores.

This paper details the new thermal-hydraulics and shutdown margin calculation modules of Studsvik Scandpower’s next-generation nodal code, SIMULATE5.

SIMULATE5’s BWR thermal-hydraulics (TH) models an entire vessel loop: core, chimney (for natural circulation reactors), upper plenum, standpipes, steam separators, down comer, re-circulation pumps, and lower plenum. The PWR thermal-hydraulics models the region from lower to upper tie plate. The core portion of the TH models of PWR and BWRs are treated essentially identically, with each assembly having an active channel and a number of parallel water channels. In each axial node of a channel, the total mixture mass, steam mass, mixture enthalpy, and mixture momentum balance equations are solved. The 3-D fuel temperatures are evaluated in the TH module by solving the radial heat conduction equation for the average pin of each node. The BWR assembly may be divided into four radial sub-channels. Assembly or nodal cross flow is allowed for PWRs.

A new three-dimensional shutdown margin (SDM) method based on “mini-core” concept has been developed in SIMULATE5. “Mini-core” geometry is used as a fast screening tool where full 3D calculations are performed for those rods identified with SDM smaller than the user’s input criteria.

Various numerical test results are presented to illustrate improvements with each module.

A new pin power reconstruction module has been implemented in Studsvik Scandpower’s next-generation nodal code, SIMULATE5. Heterogeneous pin powers are calculated by modulating multi-group pin powers from the submesh solver of SIMULATE5 with pin form factors from single-assembly CASMO5 lattice calculations. The multi-group pin power model captures instantaneous spectral effects, and actinide tracking on the assembly submesh describes exposure-induced pin power variations. Model details and verification tests against high order multi-assembly transport methods are presented. The accuracy of the new methods is also demonstrated by comparing SIMULATE5 calculations with measured critical experiment pin powers.

SIMULATE5 is a three-dimensional multi-group analytical nodal code with microscopic depletion capability. It has been developed employing “first principal models” thus avoiding ad- hoc approximations. The multi-group diffusion equations or, optionally, the simplified P3 equations are solved. Cross sections are described by a hybrid microscopic-macroscopic model that includes approximately 50 heavy nuclides and fission products. Heterogeneities in the axial direction of an assembly are treated systematically. Radially, the assembly is divided into heterogeneous submeshes, thereby overcoming the shortcomings of spatially-averaged assembly cross sections and discontinuity factors generated with zero net-current boundary conditions. Numerical tests against higher order transport methods and critical experiments show substantial improvements compared to results of existing nodal models.

Studsvik’s advanced nodal code SIMULATE3 has been in use for LWR reactor analysis for nearly 20 years. Due to limitations on memory and concerns on run time when it was developed, the macroscopic cross section model was preferred over the microscopic depletion model. The macroscopic cross-section and depletion model employs history (i.e., exposure) weighted state parameters along with tabularized two-group macroscopic cross sections. Dependencies on the history variables are determined by performing separate single assembly CASMO depletion calculations for each of the history effects. The model has been proven to be accurate for typical LWR applications. The shortcomings of the model, such as modeling the shutdown cooling effects, have been resolved with ad-hoc models.

With today’s modern desktop computers, memory and run time are no longer concerns. Also, today’s more aggressive core designs with extended cycle lengths and operation strategies require more comprehensive reactor physics tools. As the industry leader in core analysis software, Studsvik has been working on developing the next generation nodal code, SIMULATE5. As part of this development, the existing macroscopic depletion model has been replaced with a microscopic depletion model.